A. C. C. Sips

Development of ITER 15 MA ELMy H-mode inductive scenario

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.

Experimental studies of ITER demonstration discharges

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for Eaxis < 0.23–0.33 V m−1 is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps li(3) < 0.85 during the ramp up to q95 = 3. A rise phase with an H-mode transition is capable of achieving li(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q95 ~ 3 and the hybrid scenario at q95 = 4–4.5 during the FT phase is documented, providing data for the li (3) evolution after the H-mode transition and the li (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept ≤1.2 during the first half of the current decay, using a slow Ip ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

Simulations of KSTAR high performance steady state operation scenarios

We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a freeboundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; βN above 3, H98(y, 2) up to 2.0, fBS up to 0.76 and fNI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of qmin is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work. Finally, ideal MHD stability is investigated for the ITER-relevant advanced scenarios in KSTAR. The methods and results presented in this paper are expected to contribute to improving the ITER and beyond ITER predictive simulations.

Extrapolation of ASDEX Upgrade H-mode discharges to ITER

In this paper we discuss a procedure to evaluate the fusion performance of ASDEX Upgrade discharges scaled up to ITER. The kinetic profile shape is taken from the measured profiles. Multiplication factors are used to obtain a fixed Greenwald fraction and an ITER normalized thermal pressure as in the corresponding ASDEX Upgrade discharge. The toroidal field and the plasma geometry are taken from the ITER-FEAT design (scenario 2), whereas q95 is taken from the experiment. The confinement time is inferred assuming that the measured H-factor with respect to several existing scaling laws also holds for ITER. While retaining the information contained in the multi-machine databases underlying the different scaling laws, this approach adds profile effects and confinement improvement with respect to the ITER baseline, thus including recent experimental evidence such as the prediction of peaked density profiles in ITER. Under this set of assumptions, of course not unique, we estimate the ITER performance on the basis of a wide database of ASDEX Upgrade H-mode discharges, in terms of fusion power, fusion gain and triple product. According to the three scalings considered, there is a finite probability of reaching ignition, while more than half of the discharges require less auxiliary power than the one foreseen for ITER. For all the scaling laws, high values of the thermal ?N up to 2.4 are accessible. A sensitivity study gives an estimate of the accuracy of the extrapolation. The impact of different levels of tungsten concentration on the fusion performance is also studied in this paper. This scaling method is used to verify some common 0D figures of merit of ITERs fusion performance.